Evaluating techniques and phenomena of Stress Corrosion Cracking (SCC) in Light Water Reactors (LWRs); SCC evaluating techniques for predicting core internal and pipe aging of LWRs, technical data collection (Contract research)

Yamamoto, Masahiro; Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Tsujikawa, Shigeo*; Hattori, Shigeo*; Yoshii, Tsuguyasu*; et al.

JAEA-Review 2012-007, 404 Pages, 2012/03

JAEA-Review-2012-007.pdf:36.72MB

There are many LWRs which have been operated for more than 20 years in Japan and it is expected that technique corresponding to aging plants are necessary established for safety operation in LWRs. A lot of troubles related to SCC are reported and many investigations are concerned with SCC mechanism and technical evaluation. In this paper, those research data were collected as possible widely and reviewed systematically. Current circumstances concerned with SCC in LWRs were reviewed specifically as follows: SCC incidents, SCC evaluation methods for crack initiation and propagation, the investigations concerned with SCC mechanism and monitoring technique for corrosive environment. Influences with reactor types (BWR, PWR), materials (stainless steels, Ni alloys) and SCC evaluating methods (laboratories and actual plants) were summarized as graphs and tables easy to understand in common/difference points concerned with SCC. From these arranged results, future themes were considered and remarked SCC phenomenon was summarized in actual plants. As for SCC evaluations under the accelerate conditions in the laboratory test, it was suggested that a computational prediction and modeling including statistical technique and microscopic analysis in crack initiation were important. Furthermore it was suggested that monitoring techniques predicting SCC initiation and grasping plant circumstance in operation and feasibility in actual plants were important.

PIE technologies for the study of stress corrosion cracking of reactor structural materials

Ugachi, Hirokazu; Nakano, Junichi; Nemoto, Yoshiyuki; Kondo, Keietsu; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kizaki, Minoru; Omi, Masao; Shimizu, Michio

JAEA-Conf 2006-003, p.253 - 265, 2006/05

https://jopss.jaea.go.jp/pdfdata/JAEA-Conf-2006-003.pdf

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in the light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs) at hot laboratories. On the other hand, recently in the Japanese boiling water reactor (BWR) power plants, many incidents of stress corrosion cracking (SCC) of structural material such as the reactor core shrouds and primary loop recirculation (PLR) system piping were reported. In order to investigate the cause of SCC, PIEs at hot laboratories were carried out on the sample material extracted from BWR power plants. SCC studies require various kind of PIE techniques, because the SCC is caused by a complicated synergistic effects of stress and chemical environment on material that suffered degradations by irradiation and/or thermal aging. In this paper, we describe the PIE techniques adopted recently for our SCC studies, especially the crack growth measurement, uniaxial constant load (UCL) tensile test method, in-situ observation during slow strain rate test (SSRT) and several metallurgical test techniques using the FEtype transmission electron microscopy (FE-TEM), focused ion beam (FIB) processing technique, three Dimensional Atom Probe (3DAP) analysis and atomic force microscopy (AFM).

Determination of neptunium of Tokai reprocessing plant; Determination of neptunium by spectrometry in nitric acid media

Kitao, Takahiko; Nemoto, Hirokazu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Determination of Pu/Cm ratio in sludge at Tokai reprocessing plant

Okano, Masanori; Nemoto, Hirokazu*; Jitsukata, Shu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Determination of impurity in dissolved solution of spent nuclear fuel by ICP-AES

Okano, Masanori; Igarashi, Kazuto*; Nemoto, Hirokazu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Determination of uranium and acidity in reprocessing process solutions by immersion spectroscopic probe

Kuno, Takehiko; Nemoto, Hirokazu*; Okano, Masanori; Igarashi, Kazuto*; Yamada, Keiji; Watahiki, Masaru

no journal, , 

no abstracts in English

MOX reprocessing at TRP, 8; Development of hull monitor for MOX spent fuel, 1; Determination of plutonium, curium and uranium in hull pieces by destructive analysis

Suzuki, Yutaka; Kuno, Takehiko; Nemoto, Hirokazu*; Okano, Masanori; Goto, Yuichi; Igarashi, Kazuto*; Shimizu, Yasuyuki; Suda, Shizuka; Yamada, Keiji; Watahiki, Masaru

no journal, , 

no abstracts in English

Determination of trace elements in low-level radioactive liquid waste of reprocessing plant using chelating disk separation/ICP-AES

Okano, Masanori; Goto, Yuichi; Jitsukata, Shu*; Nemoto, Hirokazu*; Takano, Masato; Komoto, Norio; Kuno, Takehiko; Yamada, Keiji

no journal, , 

no abstracts in English

Determination of technetium-99 in liquid wastes of Fukushima-Daiichi Nuclear Power Plant by solid extraction-ICP AES method

Suwa, Toshio; Okano, Masanori; Jitsukata, Shu*; Nemoto, Hirokazu*; Kuno, Takehiko; Kurosawa, Akira

no journal, , 

no abstracts in English